ML21200A179

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Scale/Melcor Non-LWR Source Term Demonstration Project - High Temperature Gas-Cooled Reactor
ML21200A179
Person / Time
Site: Grand Gulf, Arkansas Nuclear, River Bend, Waterford, 07000013  Entergy icon.png
Issue date: 07/31/2021
From: Jordan Hoellman
NRC/NRR/DANU/UARP, Oak Ridge, Sandia
To:
Hoellman J
References
DE-NA0003525
Download: ML21200A179 (115)
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Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energys National Nuclear Security Administration under contract DE-NA0003525. SAND20XX-XXXX P SCALE/MELCOR Non-LWR Source Term Demonstration Project -

High-Temperature Gas-Cooled Reactor July 2021

2 NRC strategy for non-LWR source term analysis Project scope High-temperature gas-cooled reactor fission product inventory/decay heat methods and results High-temperature gas-cooled reactor plant model and source term analysis Summary Appendices SCALE overview VSOP ORIGEN library interpolation MELCOR overview MELCOR default radionuclide classes Outline

3 Integrated Action Plan (IAP) for Advanced Reactors Near-Term Implementation Action Plan Strategy 1 Knowledge, Skills, and Capacity Strategy 2 Analytical Tools Strategy 3 Flexible Review Process Strategy 4 Industry Codes and Standards Strategy 5 Technology Inclusive Issues Strategy 6 Communication ML17165A069

4 IAP Strategy 2 Volumes ML20030A177 ML20030A174 ML20030A176 ML20030A178 ML21085A484 Introduction Volume 1 Volume 2 Volume 3 Volume 4 Volume 5 ML21088A047 These Volumes outline the specific analytical tools to enable independent analysis of non-LWRs, gaps in code capabilities and data, V&V needs and code development tasks.

5 NRC strategy for non-LWR analysis (Volume 3)

6 Role of NRC severe accident codes

Project Scope

8 Understand severe accident behavior

  • Provide insights for regulatory guidance Facilitate dialogue on staffs approach for source term Demonstrate use of SCALE and MELCOR
  • Identify accident characteristics and uncertainties affecting source term
  • Develop publicly available input models for representative designs Project objectives

9 Full-plant models for three representative non-LWRs (FY21)

  • Heat pipe reactor - INL Design A
  • Pebble-bed gas-cooled reactor - PBMR-400
  • Pebble-bed molten-salt-cooled - UC Berkeley Mark I FY22
  • Molten-salt-fueled reactor - MSRE
  • Sodium-cooled fast reactor - To be determined Project scope

10

1. Build MELCOR full-plant input model Use SCALE to provide decay heat and core radionuclide inventory
2. Scenario selection
3. Perform simulations for the selected scenario and debug Base case Sensitivity cases Project approach

11 Broad Landscape High-Temperature Gas-Cooled Reactors (HTGR)

Liquid Metal Cooled Fast Reactors (LMFR)

Molten Salt Reactors (MSR)

GEH PRISM (VTR)

Advanced Reactor Concepts Westinghouse Columbia Basin Hydromine Framatome X-energy

  • StarCore General Atomics Kairos (HermeslRTR)

Terrestrial

  • Thorcon Flibe TerraPower/GEH (Natrium)*

Elysium Liquid Salt Fueled TRISO Fuel Sodium-Cooled Lead-Cooled Alpha Tech Muons Micro Reactors Oklo Stationary Transportable Ultra Safe lRTR Radiant lRTR Westinghouse (eVinci)

Liquid Salt Cooled X-energy BWX Technologies Southern (TP MCFR) lRTR Oklo ARDP Awardees MIT ACU lRTR

  • ARC-20 Demo Reactors In Licensing Review Risk Reduction Preapplication RTR Research/Test Reactor LEGEND General Atomics (EM2)

Kairos

  • TerraPower Advanced Reactor Designs

High-Temperature Gas-Cooled Reactor

13 High-temperature high-pressure helium transfers heat from core to the secondary system

  • Core outlet temperatures to 1000
  • High temperature increases efficiency
  • Fuel in a prismatic or a pebble bed core Peach Bottom Unit 1
  • Operated 1966-1974
  • 115 MW thermal power
  • 37% efficiency, 88% availability Fort St. Vrain
  • Operated 1979-1989
  • 842 MW thermal power High-temperature gas-cooled reactor (1/2)

Peach Bottom Unit #1

[https://commons.wikimedia.org/wiki/File:Peach_Bottom_-Aerial_View_1.jpg

14 Department of Energy funded design of the Next Generation Nuclear Plant

  • Project as established by Energy Policy Act of 2005
  • Project started in 2007
  • Initial focus on the PBMR-400 design
  • Pebble Bed Modular Reactor (Pty) Ltd
  • Focus of an Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) neutronics benchmark study [NEA/NSC/DOC(2013)10]
  • Areva SC-HTGR design was selected in 2012
  • Department of Energy subsequently ended support High-temperature gas-cooled reactor (2/2)

PBMR-400

[NEA/NSC/DOC(2013)10]

15 PBMR-400 - Used for SCALE/MELCOR demonstration project MELCOR model based on data from OECD/NEA neutronics benchmark project

  • Development of MELCOR Input Techniques for High Temperature Gas-cooled Reactor Analysis, James Corson, Masters thesis, Texas A&M University, 2010 No description of confinement or secondary system
  • MELCOR confinement model based on NGNP schematics
  • Simplified secondary system used to estimate steady-state conditions Publicly available design

16 400 MWt Helium coolant

  • Pressure - 9 MPa (1300 psi)
  • Core inlet - 500
  • Core outlet - 900
  • Core flowrate (downward) - 192 kg/s 452,000 TRISO pebbles in an annular core
  • Core inner diameter - 2.0 m
  • Core outer diameter - 3.7 m
  • Core height - 11 m 92 GWD/MTU target burn-up Steel vessel with graphite reflectors PBMR-400 (1/2)

Pieter J Venter, Mark N Mitchell, Fred Fortier, PBMR Reactor Design and Development, 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18), Beijing, China, August 7-12, 2005, SMiRT18-S02-2

17 TRISO particle

  • Kernel - 1.5 g U; 250 µm radius
  • Contains 14,500 TRISO particles
  • 25 mm radius
  • 5 mm graphite outer shell PBMR-400 (2/2)

TRISO particle

[INL/EXT-08-14497]

TRISO pebble

[1]

HTGR Fission Product Inventory / Decay Heat Methods & Results

19

References:

1. Status and Prospects for Gas Cooled Reactor Fuels, IAEA-TECDOC-CD-1614, April 2009
2. OECD/NEA, PBMR Coupled Neutronics / Thermal-hydraulics Transient Benchmark I: The PBMR-400 Core Design, NEA/NSC/DOC(2013)10, July 2010 PBMR-400 SCALE geometry (S. Skutnik, ORNL)

Design features

  • Fueled by graphite pebbles composed of UO2-bearing TRISO fuel particles (5-10% 235U)
  • Pebbles circulate multiple passes through the core to high discharge burnup (~90 GWd/MTIHM)

Two cases evaluated

  • Startup core: 1/3 fuel pebbles, 2/3 graphite dummy pebbles
  • Equilibrium core: 110 material zones with pre-specified material compositions (100% fuel)

PBMR-400 benchmark used to represent PBR concepts

20

  • HTR-10 initial core critical benchmark Based on International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE) benchmark for HTR-10 initial core Graphite-coated, spherical fuel elements with TRISO fuel particles

3 cm fuel spheres at 17% 235U enrichment SCALE 6.0 with ENDF/B-VIII.0 nuclear data Figure of merit: System k-eigenvalue (keff)

SCALE consistent with MCNP to within -73+/-34 pcm

MCNP and SCALE calculations both showed a moderate positive reactivity bias (1.4 +/- 0.4)%

Prior SCALE validation for HTGR systems (1/2)

G. Ilas, D. Ilas, R. P. Kelly, and E. E. Sunny, Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis, NUREG/CR-7107(ORNL/TM-2011/161), Jul. 2012 Image: NUREG/CR-7107

21

  • HTR-PROTEUS critical benchmark IRPhE benchmark based upon critical experiments performed at PROTEUS facility (Paul Scherrer Institut, Switzerland)

10 deterministic pebble packing arrangements with 3 random close-packed arrangements

Graphite-coated spherical fuel elements with TRISO fuel particles

3 cm radius graphite spheres (2.35 cm fuel region radius), 16.7% 235U enrichment Figure of merit: System k-eigenvalue (keff)

Prior SCALE validation for HTGR systems (2/2)

Difference with MCNP5 (pcm)

ENDF/B-VI ENDF/B-VII.0 Average Maximum Average Maximum Columnar hexagonal point-on-point (CHPOP) 422 +/- 93 667 +/- 82 804 +/- 87 1302 +/- 811 Hexagonal close-packed (HCP) 252 +/- 93 353 +/- 84 782 +/- 95 801 +/- 85 G. Ilas, D. Ilas, R. P. Kelly, and E. E. Sunny, Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis, NUREG/CR-7107(ORNL/TM-2011/161), Jul. 2012

22 SCALE capabilities used KENO or Shift* 3D Monte Carlo transport ENDF/B-VII.1 continuous energy physics ORIGEN for depletion Sequences

CSAS for reactivity (e.g. rod worth)

TRITON for reactor physics & depletion Relatively small amount of data except for nuclide inventory new interface file developed for inventory using standard JSON format easily read in python and post-processed into MELCOR or MACCS input contains nuclear data such as decay Q-value for traceability when performing UQ studies Workflow Power distributions Other MACCS Input MELCOR Input SCALE Binary Output Inventory Interface File SCALE Kinetics data SCALE specific Generic End-user specific SCALE Text Output

  • To be released with SCALE 6.3

23

  • ORNL has used a methodology with the Oak Ridge Isotope GENeration (ORIGEN) code to rapidly generate inventories using ORIGEN reactor libraries
  • SCALE/ORIGEN use of fundamental nuclear data allows the following to be calculated from nuclide inventory (moles of each nuclide in a system) mass decay heat activity gamma emission neutron emissions
  • With SCALE 6.2 (2016), the sequence ORIGAMI was released which is the modern approach of using ORIGEN reactor libraries General ORNL Methodology for Fuel Inventory

24

  • Soon ORIGAMI will have a new PBMR-400 fuel type and the ability to generate (in seconds) fuel inventory for a PBMR-400 pebble initial enrichment specific power history cooling time
  • Generalizing what we learn for the PBMR-400 will enable future HTGR fuel types Plans for SCALE/ORIGAMI and HTGR

>50 different fuel types supported!

Current Fuel Types

25

  • Key assumptions License applications will specify pebble circulation strategy and equilibrium core Analyzing the equilibrium core is the limiting case from an inventory/decay heat standpoint
  • Main goals Evaluate neutronic characteristics Generate inventory and decay heat for the MELCOR nodalization which may differ from how the application specifies their equilibrium core isotopics Generate individual pebble inventory within a core zone/batch (e.g., difference between fresh vs. once-through pebble in a single core zone)

Generate discharge pebble inventory/decay heat with sensitivity/uncertainty analysis HTGR analysis with SCALE: Overview PBMR-400 equilibrium core

26

1. Pebble packing
2. Temperature feedback
3. Radial/axial spectral variation
4. Pebble flow
5. TRITON model scope for ORIGEN library generation (i.e. what matters for producing one-group sections)

Analysis areas PBMR-400 equilibrium core

27 PBMR-400 benchmark specifies ~452,000 fuel pebbles with a packing fraction of 61%

Can be achieved using a BCC lattice (dodecahedral) of unbroken spheres, however substantial negative bias in keff observed due to local voids near reflector regions Present best estimate models use clipped pebbles at boundary to maintain uniform local packing fraction

  • Similar to modeling approaches used for HTR-10
1. Pebble packing J.-Y. Hong, S.-R. Wu, S.-C. Wu, D.-S. Chao, J.-H. Liang, Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre generated fuel composition library, Nuclear Engineering and Design, 374 (2021)

Image: S. Skutnik, ORNL

28 Estimation of specific reactivity feedback components (e.g., temperature reactivity coefficients of fuel, moderator) requires detailed thermal hydraulic analysis of core Strong coupling between neutronics & thermal hydraulics Approach: Using system isotherms

  • All system materials adjusted to a fixed temperature e.g., 300, 600, 900, 1200 K
  • Does not afford specific isolation of moderator / fuel temperature coefficients
2. Temperature feedback (1/2)

PBMR-400 total neutron flux, from SCALE/Shift 3D Monte Carlo Calculation (S. Skutnik, ORNL)

29

2. Temperature feedback (2/2)

Strong temperature-driven spectral shifts, especially toward 239Pu low-lying resonance Fresh core Equilibrium core

30

3. Flux shape shows a top-weighted distribution due to pebble loading & depletion Strong power peaking effects observed near graphite reflector regions (esp. interior) thermal flux

31

3. Fast : thermal flux ratio (spectral index) sensitive to radial zone; relatively invariant axially Axial Radial Reflector-adjacent (outer)

Central regions Major spectral shifts primarily occur across radial zones; i.e., primarily need radial zone Origen libraries

32 Approach: Equilibrium compositions derived from previous equilibrium core calculation with flowing pebbles (VSOP)

  • Pebble locations currently treated as static in a full-core, 3-D Monte Carlo neutron transport calculation
  • Discrete axial and radial material zones, representing spatially-dependent average at equilibrium after several months of operation Similarity to prior approaches:
  • VSOP:1 Depletion of fixed core compositions to a pre-defined keff, then shuffle zones downward, reload pebbles at top of core and repeat. Depletion assumes admixture of fresh & burned pebbles exposed to same depleting flux
  • HTR-10 multi-pass pebble burnup analysis2 follows similar procedure to VSOP
4. Continuous circulation of pebbles in the core PBMR-400 total neutron flux from SCALE/Shift 3D Monte Carlo calculation (S. Skutnik, ORNL)

References:

1.

HJ. Rütten, K.A. Haas, H. Brockmann, W. Scherer, V.S.O.P. (99/05) Computer Code System (2005) 2.

J.-Y. Hong, S.-R. Wu, S.-C. Wu, D.-S. Chao, J.-H. Liang, Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre generated fuel composition library, Nuclear Engineering and Design, 374 (2021)

33 Current assumptions:

  • Pebble transit dominated by vertical motion; can capture differential velocity across radial regions
  • Active core modeled as a right-cylindrical annulus (cylindrical shell)

Similarity to prior approaches:

VSOP: Pebble transit assumed to be in parallel vertical dimensions unless user specifies otherwise HTR-10 burnup analysis normalizes pebble residence time based on assumed transit path (conical funneling); recycled pebbles uniformly redistributed across top of core

4. Capturing possible pebble transit paths through the core (velocity differentials & cross-flow)

J.-Y. Hong, S.-R. Wu, S.-C. Wu, D.-S. Chao, J.-H. Liang, Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre generated fuel composition library, Nuclear Engineering and Design, 374 (2021)

34 Evaluate PBMR-400 cross-sections & isotopic responses at different levels of model fidelity

5. ORIGEN library analysis strategy Lower fidelity Lower computational cost High fidelity High computational cost

35

  • Accounts for important radial effects
  • Proximity to reflector
  • Effects of nearest neighbor pebbles
  • Can easily be tuned for different axial zones
5. ORIGEN library development:

reflected plane model

36

5. Plane model captures important neighbor effects Plane Pebble

37

  • Five separate cases constructed starting with a fresh pebble surrounded by non-depleting neighbors with compositions derived from PBMR-400 benchmark inventory ND-Set3
  • Pebble depleted to discharge burnup surrounded by invariant neighbors
5. ORIGEN library generation based on 5 spectral zones rzone=1 rzone=2 rzone=3 rzone=4 rzone=5

38

5. Radial, temperature effects drive differences in 1-group XSs ORIGEN libraries

39

5. Radial zone effects far more prevalent than burnup effects for pebble bed depletion Outer regions Central region Spatial-driven differences in loss cross-sections relatively stable over burnup

40 Magnitude of XS differences due to radial location increases with system temperature Gap between inner and outer regions grows with increasing temperature Implies a covariant relationship between location & temperature

5. Temperature (system isotherm) shows a large, region-dependent effect on 1G removal XS Outer regions Central region

41

  • Analysis areas
1. Pebble packing
2. Temperature feedback
3. Radial/axial spectral variation
4. Pebble flow
5. TRITON model scope for ORIGEN library generation
  • For ORIGEN library generation
  • Burnup effects appear to be second-order, roughly linear in nature
  • Radial distance from the reflector is a first-order spectrum characteristic Must be accounted for in library generation
  • Temperature (system isotherm) also a first-order effect Shows covariance with radial position Driven primarily by graphite (reflector) temperature Conclusions for pebble bed reactor ORIGEN library development Further details:

S. Skutnik, W. Wieselquist, Assessment of ORIGEN Reactor Library Development for Pebble-Bed Reactors Based on the PBMR-400 Benchmark, ORNL/TM-2020/1886, July 2021 Available on osti.gov

MELCOR High-Temperature Gas-Cooled Reactor Model

43 Fission product release

  • Release from TRISO kernel
  • Radionuclide distributions within the layers in the TRISO particle and compact
  • Release to coolant Other core models
  • Graphite oxidation
  • Intercell and intracell conduction
  • Convection & flow
  • Point kinetics
  • Dust generation and resuspension MELCOR HTGR modeling

44

  • Pebble Bed Reactor Fuel/Matrix Components Fueled part of pebble Unfueled shell (matrix) is modeled as separate component Fuel radial temperature profile for sphere
  • Prismatic Modular Reactor Fuel/Matrix Components Rod-like geometry Part of hex block associated with a fuel channel is matrix component Fuel radial temperature profile for cylinder HTGR Components Legend TRISO (FU)

Fuel (FU)

Matrix (MX)

Fluid B/C TRISO GRAPHITE Sub-component model for zonal diffusion of radionuclides through TRISO particle GRAPHITE Fuel Compact Unfueled Fueled pebble core Unfueled pebble core

45 Transient/Accident Solution Methodology Stage 1:

Normal Operation Diffusion Calculation Establish steady state distribution of radionuclides in TRISO particles and matrix Stage 2:

Normal Operation Transport Calculation Calculate steady state distribution of radionuclides and graphite dust throughout system (deposition on surfaces, convection through flow paths)

Example:

PBMR-400 Cs Distribution in Primary System Stage 3:

Accident Diffusion & Transport calculation Calculate accident progression and radionuclide release Elevation [m]

Temperature [K]

2000 K Stage 0:

Normal Operation Establish thermal state Time constant in HTGR graphite structures is very large Reduce heat capacities for structures to reach steady state thermal conditions.

Reset heat capacities after steady state is achieved.

Temperature [K]

Time [min]

Representative reflector temperature response

46 Intact TRISO Particles

  • One-dimensional finite volume diffusion equation solver for multiple zones (materials)
  • Temperature-dependent diffusion coefficients (Arrhenius form)

HTGR Radionuclide Diffusion Release Model Intact TRISO Concentrations

= 1

+

Layer FP Species Kr Cs Sr Ag D (m2/s)

Q (J/mole)

D (m2/s)

Q (J/mole)

D (m2/s)

Q (J/mole)

D (m2/s)

Q (J/mole)

Kernel (normal) 1.3E-12 126000.0 5.6-8 209000.0 2.2E-3 488000.0 6.75E-9 165000.0 Buffer 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 PyC 2.9E-8 291000.0 6.3E-8 222000.0 2.3E-6 197000.0 5.3E-9 154000.0 SiC 3.7E+1 657000.0 7.2E-14 125000.0 1.25E-9 205000.0 3.6E-9 215000.0 Matrix Carbon 6.0E-6 0.0 3.6E-4 189000.0 1.0E-2 303000.0 1.6E00 258000.0 Str. Carbon 6.0E-6 0.0 1.7E-6 149000.0 1.7E-2 268000.0 1.6E00 258000.0 Data used in the demo calculation

[IAEA TECDOC-0978]

= 0

Diffusivity Data Availability Radionuclide UO2 UCO PyC Porous Carbon SiC Matrix Graphite TRISO Overall Ag Some Not investigated Some Not found Extensive Some Extensive Cs Some Some Extensive Some Some I

Some Some Some Not found Not found Kr Some Some Not found Some Some Sr Some Some Extensive Some Some Xe Some Some Some Some Not found Iodine assumed to behave like Kr CORSOR-Booth LWR scaling used to estimate other radionuclides

47 o Recent failures - particles failing within latest time-step (burst release, diffusion release in time-step) o Previous failures - particles failing on a previous time-step (time history of diffusion release) o Contamination and recoil HTGR Radionuclide Release Models Failing Intact TRISO Released to the matrix Transition from Intact-to-failed Fuel Pebble Failed TRISO Contamination Release from failed TRISO (Modified Booth)

Intact TRISO Failed TRISO recoil recoil Released to the matrix Transfer to failed TRISO Distribution calculated from diffusion model Release from TRISO failure Diffusion Diffusion from intact TRISO Recoil fission source Fuel Pebble recoil Diffusion Diffusion

48 Steam oxidation Graphite Oxidation Reactions Air oxidation Reactions Both steam and air include rate limit due to steam/air diffusion towards active oxidation surface He H2O or Air ROX is the rate term in the parabolic oxidation equation [1/s]

49 Effective conductivity prescription for pebble bed (bed conductance)

COR Intercell Conduction

  • Tanaka and Chisaka expression for effective radial conductivity (of a single PMR hex block)
  • A radiation term is incorporated in parallel with the pore conductivity
  • Thermal resistance of helium gaps between hex block fuel elements is added in parallel via a gap conductance term Effective conductivity prescription for prismatic (continuous solid with pores)
  • Zehner-Schlunder-Bauer with Breitbach-Barthels modification to the radiation term Dp=.06 m Kf=.154 W/m-K Ks = 26 W/m-K Ks = 26 W/m-K Kf=.154 W/m-K

50 Heat transfer coefficient (Nusselt number) correlations for pebble bed convection:

  • Isolated, spherical particles
  • Use Tfilm to evaluate non-dimensional numbers, use maximum of forced and free Nu
  • Constants and exponents accessible by sensitivity coefficient Interface Between Thermal-hydraulics and Pebble Bed Reactor Core Structures Flow resistance
  • Packed bed pressure drop

, = 1 + 2 1

+ 3 1

4 1

Loss coefficient relative to Ergun (original) coefficient at Re=1000

= 2.0 + 0.6

1 4

1 3

= 2.0 + 0.6

1 2

1 3

51 Standard treatment Feedback models

  • User-specified external input
  • Doppler
  • Fuel and moderator density Point kinetics modeling

=

+

=1 6

+ 0

=

= 1 6

High-Temperature Gas-Cooled Reactor Plant Model and Source Term Analysis

53 Reactor vessel and core

[P.J. Venter, M.N. Mitchell, F. Fortier, PBMR reactor design and development, in: Proceedings from the 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18), Beijing, China, Aug. 2005]

171 165 125 135 155 145 170 170 Core Radial Ring 1

3 4

5 6

7 8

2 161 142 122 132 152 121 131 151 141 221 241 231 251 162 160 141 121 131 151 120 130 150 140 221 241 231 251 161 162 143 123 133 1

153 122 132 152 142 163 163 144 124 134 154 123 133 153 143 164 164 145 125 135 155 124 134 154 144 165 112 111 110 113 114 115 181 Inlet Riser 182 146 126 136 156 166 6

176 186 100 304 303 302 301 Vessel control volume, flow path, and heat structure nodalization with core package boundaries in blue Vessel core package nodalization (8 rings x 29 axial levels)

Correct aspect ratio Pebble bed Rings 2-6, Axial levels 6-27

54 Reactor building HTGR Mechanistic Source Terms White Paper, July 2010, [INL-EXT-10-17999]

Nodalization guided by NGNP layout

  • Passive air-flow Reactor Cavity Cooling System (RCCS)

Leakage assumed to be the same as BWR Mark I reactor building surrounding the containment

  • 100% vol/day at 0.25 psig Picture above shows a water-cooled RCCS but demo model uses air-cooled RCCS.

55 Recirculation loop and secondary heat removal 615 605 605 Hot leg 600 610 600 640 610 620 620 630 630 Cold leg Vessel outlet plenum 110 Vessel inlet riser 181 500 500 510 Primary heat exchanger 626 625 627 Recirculation system and secondary heat removal

  • Recirculation loop and secondary heat removal provide boundary conditions to the vessel Flow rate Heat removal & inlet temperature Pipe break nodalization allows counter-current natural circulation flow
  • MELCOR counter-current flow model used to represent adjacent stream drag forces
  • Geometry similar to PWR hot leg natural circulation [NUREG-1922]
  • Allows for air ingression Scenario: depressurized loss of forced circulation (DLOFC)
  • Assumes double-ended break of the hot leg

56 DLOFC is initiated after 900 days of operation

  • Long-term fission product concentrations developed in TRISO and pebble
  • 24 kg/yr graphite dust generation based on German AVR experience
  • TRISO initialized with 10-5 failure fraction during the steady state Provisions for air ingression Reactor cavity cooling system (RCCS) is operational Individual sensitivity calculations to explore variations in the model response to uncertainty in input parameters DLOFC scenario

57 0

1 2

3 4

5 6

7 8

9 10 0

0.2 0.4 0.6 0.8 1

Pressure (MPa)

Time (min)

Vessel Pressure DLOFC reference case results (1/7) 9.3 MPa initial pressure Following pipe break

  • The vessel depressurizes in seconds as the high-pressure helium escapes out both sides of the broken pipe
  • Peak velocity in the pebble bed is 45 m/s (normal flow rate is 11-18 m/s)

Counter-current flow established on the vessel side of the pipe break

  • Hot gases from the exit plenum escape on the top side of the broken hot leg pipe and cooler gases enter along the bottom of the pipe

-0.1

-0.08

-0.06

-0.04

-0.02 0

0.02 0.04 0.06 0.08 0.1 0

24 48 72 Flowrate (kg/s)

Time (hr)

Pipe Break Flowrates Vessel side lower Vessel side upper SG side Negative flow is into the reactor Positive flow is out of the reactor

58 DLOFC reference case results (2/7)

-0.100

-0.075

-0.050

-0.025 0.000 0.025 0.050 0.075 0.100 0

24 48 72 Velocity (m/s)

Time (hr)

Axial flow velocities in outer region of the core Axial flow velocities in inner region of the core Positive flow is downward Negative flow is upward 0

200 400 600 800 1000 1200 1400 1600 1800 0

24 48 72 Temperature (deg-C)

Time (hr)

Top (inlet) is coolest when helium circulator is operating Exit hottest at the bottom of the pebble bed Exit is cooler due to the flow reversal and cooler air entering from the exit plenum In-vessel natural circulation flow after blowdown

  • Upward flow in the inner region of the core where the fuel temperatures and decay power heating are higher
  • Downward flow in the outer region of the core where the fuel temperatures and decay power heating are lower
  • Flow increases when the fuel starts to cool The fuel temperatures in the inner region of the pebble bed shift from cooler at inlet and hot at the outlet due to the flow reversal
  • The axial fuel temperatures are affected by the local decay heat power (highest in the center) and the flow direction During normal operation, the fuel at the exit (bottom) is the hottest The exit becomes the coolest location (low power and cooler gases entering from the exit plenum)

59 0

1 2

3 4

5 0

24 48 72 96 120 144 168 Power (MW)

Time (hr)

Decay Heat Power Decay Heat Oxidation 0

100 200 300 400 500 600 700 800 0

24 48 72 Mass (kg)

Time (hr)

Graphite Oxdidation By-products CO CO2 The core heatup is dominated by the decay heat

  • The air oxidation power is relatively small at <25 kW
  • Although the vessel is thermally-stratified with a low exit path, a small natural circulation flow persists to bring air into the vessel Pebble bed inlet and circulation velocities are <0.04 m/s The graphite oxidation produces significant quantities of CO and CO2
  • Approximately 50% of the oxidation occurs in the graphite reflector structures around the inlet plenum and 50% in the lower portion of the pebble bed.
  • ~1% of the pebble matrix oxidized after 168 hr 17% peak pebble oxidation at the bottom center DLOFC reference case results (3/7)

Maximum oxidation power is <25 kW

60 0

0.2 0.4 0.6 0.8 1

0 24 48 72 Mole Fraction (-)

Time (hr)

CO Reactor Building Mole Fraction DLOFC reference case results (4/7)

Potential for combustion in the reactor building

  • MELCOR lower limit for CO combustion with an ignition source is 12.9% (~2X higher than for hydrogen)
  • Highly dependent on local concentrations and building design and interconnectivity
  • Demo reactor building assumes high inter-connectivity Allows air and CO circulation
  • No carbon-dioxide burns were predicted through 168 hr Reactor building CO concentration Lower CO flammability limit

61 MELCOR predicts release and transport from fuel to the environment

  • Fuel heat-up
  • TRISO layers - Initial failure fraction + failures during heat-up
  • Pebble matrix and pebble outer shell - Higher diffusivity at elevated temperatures, recoil, and air oxidation
  • Primary system - Failed with the initiating event
  • Reactor building - Design leakage DLOFC reference case results (5/7) 0.00001 0.0001 0.001 0.01 0.1 1

1200 1400 1600 1800 2000 2200 2400 2600 Failure Fraction (-)

Temperature (°C)

TRISO Failure Fraction vs Temperature Specified failure versus temperature 0.00001 0.0001 0.001 0.01 0.1 1

0 24 48 72 96 120 144 168 Failure Fraction (-)

Time (hr)

TRISO Failure Fraction

~2x10-4 failure fraction calculated in the reference case

62 1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 0

24 48 72 Fraction of initial inventory (-)

Time (hr)

Cesium Distribution Release In-vessel Reactor building Environment DLOFC reference case results (6/7)

The impact of the low TRISO failure fraction leads to small releases

  • Larger cesium release due its the higher diffusivity
  • Ag release to the environment is 1.2x10-3 (highest diffusivity) 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 0

24 48 72 Fraction of initial inventory (-)

Time (hr)

Iodine Distribution Release In-vessel Reactor building Environment Initial release dominated by airborne dust from steady operation Initial release dominated by airborne dust from steady operation

63 Of the small release from the fuel 34% and 62% of iodine and cesium, respectively, retained in the vessel

  • Thermally-stratified orientation limits vessel releases
  • Low flowrate combined with aerosol deposition
  • Inclusion of graphite oxidation reaction products (CO and CO2) promotes more flow and therefore more releases from the vessel 58% and 34% of iodine and cesium, respectively, retained in the reactor building
  • No strong driving force for reactor building leakage Reference model uses a hole size equivalent to 100%

leakage per day at a design pressure of 0.25 psig (3.2 in2)

DLOFC reference case results (7/7) 33.6%

57.5%

8.9%

Iodine Distribution at 7 days In-vessel Reactor building Environment 61.7%

34.4%

3.9%

Cesium Distribution at 7 days In-vessel Reactor building Environment

64 MELCOR can be used to explore the variability of the results to uncertainties Model Parameter Distribution Range TRISO Model Parameters Initial TRISO Failure Fraction (fraction of inventory)

Log uniform 10 10-3 TRISO Failure Rate Multiplier (-)

Log uniform 0.1 - 10.0 Intact TRISO Diffusivity Multiplier (-)

Log uniform 0.001 - 1000.0 Failed TRISO Diffusivity Multiplier (-)

Log uniform 0.001 - 1000.0 Matrix Diffusivity Multiplier (-)

Log uniform 0.001 - 1000.0 TRISO Pebble Emissivity (-)

Uniform 0.5 - 0.999 TRISO Pebble Bed Porosity (-)

Uniform 0.3 - 0.5 TRISO recoil fraction (-)

Uniform 0 - 0.03 Radionuclide Model Parameters Shape Factor (-)

Uniform 1.0 - 5.0 Gaseous Iodine Multiplier (Base = 5% I2)

Uniform 0.02 - 1.0 Design Parameters Graphite Conductivity Multiplier (-)

Uniform 0.5 - 1.5 Decay Heat Multiplier (-)

Uniform 0.9 - 1.1 RCCS Blockage Multiplier (-)

Log uniform 0.001 - 1.0 RCCS Emissivity (-)

Uniform 0.1 - 1.0 Reactor Building Leakage Multiplier (-)

Log uniform 0.1 - 100.0 Wind speed (m/s)

Uniform 0 - 10

65 Single parameter sensitivity results (1/4)

The sensitivity parameters were sampled at the minimum and maximum values to illustrate their impacts

  • A low graphite conductivity has the largest impact on the peak fuel temperature Graphite conductivity varies considerably with irradiation

(>10X) and also varies with temperature

  • +/-10% decay heat has next largest impact on the peak fuel temperature
  • High/low emissivity, the next most important single factor, is used as a surrogate for the relative importance of radiative exchange in the pebble bed
  • Debris bed porosity had a small effect on the peak fuel temperature
  • Heat dissipation limits the magnitude of the initial peak for a blocked RCCS Slow heat-up to 1800by 7 days 200 400 600 800 1000 1200 1400 1600 1800 2000 0

24 48 72 96 120 144 168 Temperature (°C)

Time (hr)

Peak Fuel Temperature Base case Low Gr k High Gr k 1.1X decay heat 0.9X decay heat Low pebble emissivity High pebble emissivity Low pebble porosity High pebble porosity Blocked RCCS

66 Examples of single parameter sensitivity results (2/4)

As the peak fuel temperature rises, the TRISO failure fraction increases

  • Blocked RCCS does not have impact for several days The cesium environmental release shows an order of magnitude variation
  • Reflects variations in release from the pebbles
  • Graphite conductivity had the largest impact
  • Variations in emissivity = uncertainty in radiative heat transport (similar to +/-10% in decay heat power)
  • Pebble porosity had a small impact 0.00001 0.0001 0.001 0.01 0.1 0

24 48 72 96 120 144 168 Failure Fraction (-)

Time (hr)

TRISO Failure Fraction Base case Low Gr k High Gr k 1.1X decay heat 0.9X decay heat Low pebble emissivity High pebble emissivity Low pebble porosity High pebble porosity Blocked RCCS 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0

24 48 72 96 120 144 168 Release Fraction (-)

Time (hr)

Cesium Release Fraction from the Pebbles Base case Low Gr k High Gr k 1.1X decay heat 0.9X decay heat Low pebble emissivity High pebble emissivity Low pebble porosity High pebble porosity 2x recoil No recoil Blocked RCCS

67 Examples of single parameter sensitivity results (3/4)

Larger hole size in the building and higher wind speed causes higher releases to environment

  • 100X building leakage has less than a 10X impact
  • External wind has small effect Graphite oxidation and the associated CO/CO2 production did not increase the source term
  • CO/CO2 gas production did not increase environment release Early impacts of the recoil and initial TRISO failure fraction did not impact long-term environmental release
  • Magnitude of the release dominated by the fuel temperature response and the TRISO failure model Late step change in the blocked RCCS release is due to a carbon monoxide burn
  • Building pressurization forces out airborne radionuclides 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0

24 48 72 96 120 144 168 Release Fraction (-)

Time (hr)

Cesium Environmental Release Base 1X RB Leakage, 10 m/s 10X RB leakage, 0 m/s wind 10X RB leakage, 10 m/s wind 100X RB leakage, 0 m/s wind 100X RB leakage, 10 m/s wind 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0

24 48 72 96 120 144 168 Release Fraction (-)

Time (hr)

Cesium Environmental Release Base case No oxidation 2X Recoil No Recoil 10X Initial TRISO failure 0.1X Initial TRISO failure Blocked RCCS

68 Examples of single parameter sensitivity results (4/4)

Blocked RCCS leads to higher CO generation

  • Ratio of reaction products is dependent on the temperature of the graphite
  • Blocked RCCS generates ~9% more moles of CO and CO2 Higher CO generation led to a burn in the steam generator compartment (pipe break location)
  • Incomplete burn with slow flame speed Low oxygen concentration (6.8%)
  • 0.25 bar (3.5 psi) pressure rise
  • Burn creates non-condensable CO2 No subsequent condensation 0

250 500 750 1000 1250 1500 1750 0

24 48 72 96 120 144 168 Mass (kg)

Time (hr)

Graphite Oxdidation By-products CO CO2 CO - Blocked RCCS CO2 - Blocked RCCS 0

0.1 0.2 0.3 0.4 0.5 0

24 48 72 96 120 144 168 Mole Fraction (-)

Time (hr)

Reactor Building CO Mole Fraction Base Blocked RCCS CO Lower Flammability Limit CO burn in the SG compartment

High-Temperature Gas-Cooled Reactor Uncertainty Analysis

Role of MELCOR in Resolving Uncertainty Simulation Uncertainty Plant Initial/Boundary Condition Uncertainty Event Scenario Uncertainty Phenomenological Model Uncertainty SSC Failure Modes Uncertainty Engineering Performance Risk-Informed Assessment

71 Overall motivation

  • A clustering of system responses provides insights on important assumptions and modeling parameters
  • Provides a most likely release and range of releases for the scenario MELCOR application to LWRs
  • Range of SOARCA uncertainty studies
  • PWR and BWR plant uncertainty studies
  • Resolved role of uncertainty in critical severe accident issues Commonalities between LWR and HTGR
  • Chemical form of key elements
  • Aerosol physics parameters (e.g., shape factor)
  • Operating time before accident happens
  • Containment leakage hole size Parameter selection emphasized potential HTGR-specific uncertainties
  • Ran 2000 realizations on High Performance Computer Evolution from MELCOR LWR Uncertainty Analysis

72 Parametric Uncertainty - Capability Demonstration Model Parameter Distribution Range TRISO Model Parameters Initial TRISO Failure Fraction (fraction of inventory)

Log uniform 10 10-3 TRISO Failure Rate Multiplier (-)

Log uniform 0.1 - 100.0 Intact TRISO Diffusivity Multiplier (-)

Log uniform 0.001 - 1000.0 Failed TRISO Diffusivity Multiplier (-)

Log uniform 0.001 - 1000.0 Matrix Diffusivity Multiplier (-)

Log uniform 0.001 - 1000.0 TRISO Pebble Emissivity (-)

Uniform 0.5 - 0.999 TRISO Pebble Bed Porosity (-)

Uniform 0.3 - 0.5 TRISO recoil fraction (-)

Uniform 0 - 0.03 Radionuclide Model Parameters Shape Factor (-)

Uniform 1.0 - 5.0 Gaseous Iodine Multiplier (Base = 5% I2)

Uniform 0.02 - 1.0 Design Parameters Graphite Conductivity Multiplier (-)

Uniform 0.5 - 1.5 Decay Heat Multiplier (-)

Uniform 0.9 - 1.1 RCCS Blockage Multiplier (-)

Log uniform 0.001 - 1.0 RCCS Emissivity (-)

Uniform 0.1 - 1.0 Reactor Building Leakage Multiplier (-)

Log uniform 0.1 - 100.0 Wind speed (m/s)

Uniform 0 - 10

73 UO2 Thermal Response

74 UO2 Thermal Transient Evolution Core cells with peak fuel temperatures at end of simulation Simulation time denoted as accident phase These core cells do not exhibit cooldown prior to end of accident phase

75 TRISO Particle Failure Initial distribution of failed TRISO particles Long-term TRISO particle failure possible for core cells exhibiting prolonged over-temperatures

76 Evolution of TRISO Particle Failures Tails of realizations contributing to longer term growth of TRISO particle failures 50th percentile reasonably stable in the long-term Rapid growth in failure fraction driven by the early temperature excursion Long-term failures of TRISO particles at lower rate but driven by prolonged period of elevated fuel temperature Lower rates of failure entirely driven by early temperature excursion Variability in peak fuel temperature and cooldown transient dominates higher failure rate realizations

77 Role of Decay Heat Rejection - Latest Time to Peak Fuel Temperature

78 Role of Decay Heat Rejection - Peak Fuel Temperature

Summary

80 Conclusions Added HTGR modeling capabilities to SCALE & MELCOR for HTGR source term analysis to show code readiness Modeling demonstrated for a DLOFC Scenario

  • Input of detailed ORIGEN radionuclide inventory data from ORNL
  • Input radial and axial power distributions from ORNL neutronic analysis
  • Develop MELCOR input model for exploratory analysis
  • Fast-running calculations facilitate sensitivity evaluations Developed an understanding of non-LWR beyond-design-basis-accident behavior and overall plant response

SCALE Overview

82 SCALE Development for Regulatory Applications What Is It?

The SCALE code system is a modeling and simulation suite for nuclear safety analysis and design. It is a modernized code with a long history of application in the regulatory process.

How Is It Used?

SCALE is used to support licensing activities in NRR (e.g., analysis of spent fuel pool criticality, generating nuclear physics and decay heat parameters for design basis accident analysis) and NMSS (e.g., review of consolidated interim storage facilities, burnup credit).

Who Uses It?

SCALE is used by the U.S. Nuclear Regulatory Commission (NRC) and in 61 countries (about 10,000 users and 33 regulatory bodies).

How Has It Been Assessed?

SCALE has been validated against criticality benchmarks (>1000), destructive assay of fuel and decay heat for PWRs and BWRs (>200)

83 Data to generate for MELCOR: QOIs

VSOP Backup Slides

85 Fuel shuffling /

pebble recycle VSOP workflow shares several features of conventional 2-step LWR core analyses Core inventories Single-element /

assembly flux solution Simplified transport /

diffusion

  • Homogenized material regions with few-group cross-sections Few-group cross-sections (critical spectrum)

Spatial flux / power distribution across the core Depletion update (pin / pebble)

Region-wise flux solution Updated material inventories

86 VSOP calculation flow (MEDUL)

Neutron spectrum Neutron diffusion (2-D / 3-D)

Burnup / depletion Thermal hydraulics keff ktarget ?

Material update (shuffle)

BU BUmax ?

Disposal storage Ex-core decay storage Core material regions Loaded material region Fresh fuel pebbles Discharged region Discharged batch YES NO NO YES

87 VSOP calculation flow (MEDUL)

Neutron spectrum Neutron diffusion (2-D / 3-D)

Burnup / depletion Thermal hydraulics keff ktarget ?

Material update (shuffle)

BU BUmax ?

Disposal storage Ex-core decay storage Core material regions Loaded material region Fresh fuel pebbles Discharged region Discharged batch YES NO NO YES This is just two-step neutronics (Polaris+PARCS)

88 VSOP calculation flow (MEDUL)

Neutron spectrum Neutron diffusion (2-D / 3-D)

Burnup / depletion Thermal hydraulics keff ktarget ?

Material update (shuffle)

BU BUmax ?

Disposal storage Ex-core decay storage Core material regions Loaded material region Fresh fuel pebbles Discharged region Discharged batch YES NO NO YES At equilibrium, spatial distributions are static: power, neutron spectrum, isotopics!

We use Monte Carlo to generate the high-fidelity spatial flux spectrum and one-group cross sections

89 VSOP calculation flow (MEDUL)

Neutron spectrum Neutron diffusion (2-D / 3-D)

Burnup / depletion Thermal hydraulics keff ktarget ?

Material update (shuffle)

BU BUmax ?

Disposal storage Ex-core decay storage Core material regions Loaded material region Fresh fuel pebbles Discharged region Discharged batch YES NO NO YES We simulate a pebble* moving through the equilibrium core with a time-dependent power and flux spectrum based on its position.

This pebble* can be used to reconstruct the detailed core composition or iterate on the equilibrium core.

  • equivalent to a batch of pebbles with same history

90 Iterative procedure for developing equilibrium core compositions Determine average burnup of each pebble batch within a zone (axial / radial)

Deplete each batch within zone to its respective burnup

  • Origen library based on region-wise flux from core transport Average zone compositions
  • Weighted sum of batches Calculate core power distribution & flux shape by zone
  • Generate ORIGEN library for each zone Repeat on initial guess inventories until keff converges; depleted compositions represent approximate equilibrium

91

  • Were interested in determining equilibrium compositions and flux shape by region Not trying to perform dynamics or reload analysis; just need equilibrium in-core inventories
  • At-equilibrium assumption simplifies analysis Conservative and bounding: i.e., converged upon highest core-averaged burnup (and thus highest fission product inventories) 2-step analysis requires many repeated calculations

e.g., 22 axial zones x 5 passes through core => 110 calculations to perform one complete cycle! (Still not at equilibrium)

Feasible with few-group diffusion, costly for MG transport!

Why use an iterative approach to equilibrium core compositions (instead of 2-step?)

ORIGEN Library Interpolation Backup Slides

93

  • Rapid answers to common questions such as What I/Cs/Pu content could I expect in a PBMR-400 pebble at 90 GWd/MTU?

a.

assuming constant power?

b.

pass-dependent power?

c.

during a power maneuver?

d.

after 4 days of decay?

e.

after 40 days of decay?

f.

after 40 years of decay?

g.

at 80 GWd/MTU?

h.

in a pebble with +1% enrichment?

  • Up-front work required Sensitivity analysis of the reactor system to understand the state changes that impact neutron flux spectrum in the fuel (e.g. moderator density in BWR)

Running many CPU-hours of TRITON coupled transport+depletion cases to generate a database of 1-group cross sections which can be interpolated to a specific state (ORIGEN reactor library)

Those libraries can then be used later (in ORIGAMI) to regenerate inventory and reaction rates:

() = () () ()

Aspects of the ORNL methodology for fuel inventory Each answer requires a <10 second calc. on a single CPU Why is speed important? This approach is not just for seeding MELCOR nodalizations. All back-end analysis can use this approach: dry storage casks, on-site storage, discharge inventory analysis, transportation packages.

Why do it this way?

If is insensitive to decay time, power level, then b through h can be answered from a single TRITON pre-calculation!

94 What level of TRITON model fidelity is required to generate a reasonable 1-group xs database (ORIGEN reactor library) for rapid LWR inventory calculations?

a.

3D full-core with plant-specific loading pattern b.

3D full-core with equilibrium loading pattern c.

3D core subset d.

3D single assembly e.

2D core subset f.

2D single assembly g.

2D single pin h.

0D infinitely homogeneous mixture For LWRs, using 2D single assembly models to generate the 1-group xs database appears sufficient!

verification confirms ORIGAMI reproduces TRITON results with same (simple) operating history validation against spent fuel inventory and decay heat measurements confirms the overall approach is adequate code results generally within experimental uncertainty bands

<1% error in decay heat, <5% error in important nuclides, <15% error in others Strategy for LWRs Requires plant-specific knowledge Assembly position matters Imposes additional assumptions or requires too much information!

Has trouble with local variations (control elements, water holes, channel box)

Increasing fidelity Has trouble if any geometry is important

95 What level of TRITON model fidelity is required to generate a reasonable 1-group xs database for rapid HTGR inventory calculations?

a.

3D full-core with plant-specific pebble loading & discharge strategy b.

3D full-core with equilibrium pebble distribution c.

2D core slice with equilibrium pebble distribution d.

1D single pebble with buffer for neighbor effects e.

1D single pebble f.

0D infinitely homogeneous mixture Using at SCALE/TRITON 3D full-core at equilibrium (b) is equivalent to VSOP but with:

ENDF/B-VII.1+ modern nuclear data SCALE complete ORIGEN nuclide set instead of VSOP limited set SCALE high-fidelity full-core Monte Carlo transport instead of VSOP diffusion Strategy for HTGRs Requires plant-specific knowledge Previously investigated in other work; difficult to optimize buffer Does not account for reflectors Used in this study to understand sensitivity to model fidelity Computationally expensive

96

  • First, understand the state changes that influence the neutron flux spectrum in a pebble as it flows through an equilibrium core:

a.

pebble power history b.

pebble burnup c.

axial position in the core d.

radial position in the core (proximity to radial reflector) e.

pebble neighbors (burnup/temperature/inventory) f.

temperature Next, generalize the SCALE concept of the ORIGEN reactor library for HTGR / PBMR-400 Our focus for the PBMR-400

97 pr = [ pr1 pr2... pr ]

pz = [ p1 p2 pn ]

ztime = [ rt1 rt2 rtn ]

hist[

pass{ power=180 burn=64 down=7 rzone=ANY }

pass{ power=160 burn=62 down=6 rzone=ANY }

pass{ power=140 burn=64 down=7 rzone=3 }

]

Prototype ORIGAMI input for multi-pass pebble inventory calculations (SCALE 7.0) radial power shape axial power shape (relative) residence time in each axial zone Example history: 3-pass pebble history, each pass moves through declared axial zones power: average MWd/MTU for that pass burn: days at power down: days decay rzone: radial zone ORIGAMI operating history input

98

  • Legacy ORIGEN library interpolation (via ARP) optimized for LWR analysis Interpolation dimensions of initial enrichment, average moderator density, burnup
  • Diverse physics characteristics of non-LWR cores require new dimensions for reactor library interpolation e.g., PBMR: radial distance from reflector, initial pebble enrichment, reflector temperature
  • To address this, we have developed a new HDF5-based format for self-describing ORIGEN libraries capable of accommodating arbitrary dimensions for interpolation Enhancing ORIGEN library interpolation capabilities to accommodate non-LWR systems

99 Legacy ORIGEN reactor data library interpolation relies on an ASCII database with hard-coded interpolation dimensions 99 Fuel Cycle Scenario Modeling Workshop - CyBORG arpdata.txt lib1 lib2 lib4 lib5 lib3 lib6 Assembly1 lib1 lib2 lib4 lib5 lib3 lib6 Assembly2 lib1 lib2 lib4 lib5 lib3 lib6 Assembly3 Pre-defined dimensions:

  • 235U Enrichment
  • Moderator density Individual permutations on interpolation dimensions

100 New HDF5-based Archive format designed to accommodate arbitrary interpolation dimensions TransitionStructure Lib #1 Lib #2 Tags DecayData Loss XS Fission XS Neutron yields Energy per fission Energy per capture Transition matrix particle neutron Lib #3 Tags Tags HDF5 Archive Initial enrichment Refueling rate FP removal rate

MELCOR for Accident Progression and Source Term Analysis

102 MELCOR Development for Regulatory Applications What Is It?

MELCOR is an engineering-level code that simulates the response of the reactor core, primary coolant system, containment, and surrounding buildings to a severe accident.

Who Uses It?

MELCOR is used by domestic universities and national laboratories, and international organizations in around 30 countries. It is distributed as part of NRCs Cooperative Severe Accident Research Program (CSARP).

How Is It Used?

MELCOR is used to support severe accident and source term activities at NRC, including the development of regulatory source terms for LWRs, analysis of success criteria for probabilistic risk assessment models, site risk studies, and forensic analysis of the Fukushima accident.

How Has It Been Assessed?

MELCOR has been validated against numerous international standard problems, benchmarks, separate effects (e.g., VERCORS) and integral experiments (e.g., Phebus FPT), and reactor accidents (e.g., TMI-2, Fukushima).

103 Source Term Development Process Fission Product Transport MELCOR Oxidation/Gas Generation Experimental Basis Melt Progression Fission Product Release PIRT process Accident Analysis Design Basis Source Term Scenario # 1 Scenario # 2 Synthesize timings and release fractions Cs Diffusivity Scenario # n-1 Scenario # n

104 SCALE/MELCOR/MACCS Safety/Risk Assessment

  • Technology-neutral o

Experimental o

Naval o

Advanced LWRs o

Advanced Non-LWRs

  • Accident forensics (Fukushima, TMI)
  • License amendments
  • Risk-informed regulation
  • Design certification (e.g.,

NuScale)

  • Vulnerability studies
  • Emergency Planning Zone Analysis Design/Operational Support
  • Design analysis scoping calculations
  • Training simulators Fusion
  • Neutron beam injectors
  • ITER cryostat modeling
  • He-cooled pebble test blanket (H3)

Spent Fuel

  • Risk studies
  • Multi-unit accidents
  • Dry storage
  • Spent fuel transport/package applications Facility Safety
  • Leak path factor calculations
  • DOE safety toolbox codes
  • DOE nuclear facilities (Pantex, Hanford, Los Alamos, Savannah River Site)

Nuclear Reactor System Applications Non-Reactor Applications SCALE Neutronics

  • Criticality
  • Shielding
  • Radionuclide inventory
  • Burnup credit
  • Decay heat MELCOR Integrated Severe Accident Progression
  • Hydrodynamics for range of working fluids
  • Accident response of plant structures, systems and components
  • Fission product transport MACCS Radiological Consequences
  • Near-and far-field atmospheric transport and deposition
  • Assessment of health and economic impacts

105 Phenomena modeled Fully integrated, engineering-level code

  • Thermal-hydraulic response of reactor coolant system, reactor cavity, rector enclosures, and auxiliary buildings
  • Core heat-up, degradation and relocation
  • Core-concrete interaction
  • Flammable gas production, transport and combustion
  • Fission product release and transport behavior Level of physics modeling consistent with
  • State-of-knowledge
  • Necessity to capture global plant response
  • Reduced-order and correlation-based modeling often most valuable to link plant physical conditions to evolution of severe accident and fission product release/transport Traditional application
  • Models constructed by user from basic components (control volumes, flow paths and heat structures)
  • Demonstrated adaptability to new reactor designs - HPR, HTGR, SMR, MSR, ATR, Naval Reactors, VVER, SFP, MELCOR Attributes Foundations of MELCOR Development

106 Validated physical models

  • International Standard Problems, benchmarks, experiments, and reactor accidents
  • Beyond design basis validation will always be limited by model uncertainty that arises when extrapolated to reactor-scale Cooperative Severe Accident Research Program (CSARP) is an NRC-sponsored international, collaborative community supporting the validation of MELCOR International LWR fleet relies on safety assessments performed with the MELCOR code MELCOR Attributes MELCOR Pedigree International Collaboration Cooperative Severe Accident Research Program (CSARP) - June/U.S.A MELCOR Code Assessment Program (MCAP) - June/U.S.A European MELCOR User Group (EMUG) Meeting - Spring/Europe European MELCOR User Group (EMUG) Meeting - Fall/Asia

107 Common Phenomenology

108 Modeling is mechanistic consistent with level of knowledge of phenomena supported by experiments Parametric models enable uncertainties to be characterized Majority of modeling parameters can be varied Properties of materials, correlation coefficients, numerical controls/tolerances, etc.

Code models are general and flexible Relatively easy to model novel designs All-purpose thermal hydraulic and aerosol transport code MELCOR Modeling Approach

MELCOR State-of-the-Art MELCOR Code Development M2x Official Code Releases Version Date 2.2.18180 December 2020 2.2.14959 October 2019 2.2.11932 November 2018 2.2.9541 February 2017 2.1.6342 October 2014 2.1.4803 September 2012 2.1.3649 November 2011 2.1.3096 August 2011 2.1.YT August 2008 2.0 (beta)

Sept 2006

110 MELCOR Software Quality Assurance - Best Practices MELCOR Wiki

  • Archiving information
  • Sharing resources (policies, conventions, information, progress) among the development team.

Code Configuration Management (CM)

  • Subversion
  • TortoiseSVN
  • VisualSVN integrates with Visual Studio (IDE)

Reviews

  • Code Reviews: Code Collaborator
  • Internal SQA reviews Continuous builds & testing
  • DEF application used to launch multiple jobs and collect results
  • Regression test report
  • More thorough testing for code release
  • Target bug fixes and new models for testing Emphasis is on Automation Affordable solutions Consistent solutions MELCOR SQA Standards SNL Corporate procedure IM100.3.5 CMMI-4+

NRC NUREG/BR-0167 Bug tracking and reporting Bugzilla online Code Validation Assessment calculations Code cross walks for complex phenomena where data does not exist.

Documentation Available on Subversion repository with links from wiki Latest PDF with bookmarks automatically generated from word documents under Subversion control Links on MELCOR wiki Project Management Jira for tracking progress/issues Can be viewable externally by stakeholders Sharing of information with users External web page MELCOR workshops MELCOR User Groups (EMUG & AMUG)

111 MELCOR Verification & Validation Basis AB-1 AB-5 T-3 Sodium Fires (Completed)

Molten Salt (planned)

Air-Ingress Helical SG HT MSRE experiments HTGR (planned)

Sodium Reactors (planned)

LOF,LOHS,TOP TREAT M-Series ANL-ART-38 Volume 1: Primer & User Guide Volume 2: Reference Manual Volume 3: MELCOR Assessment Problems Analytical Problems Saturated Liquid Depressurization Adiabatic Expansion of Hydrogen Transient Heat Flow in a Semi-Infinite Heat Slab Cooling of Heat Structures in a Fluid Radial Heat Conduction in Annular Structures Establishment of Flow Specific to non-LWR application LWR & non-LWR applications

[SAND2015-6693 R]

112 Sample Validation Cases Case 1a 1b 2a 2b 3a 3b US/INL 0.467 1.0 0.026 0.996 1.32E-4 0.208 US/GA 0.453 0.97 0.006 0.968 7.33E-3 1.00 US/SNL 0.465 1.0 0.026 0.995 1.00E-4 0.208 US/NRC 0.463 1.0 0.026 0.989 1.25E-4 0.207 France 0.472 1.0 0.028 0.995 6.59E-5 0.207 Korea 0.473 1.0 0.029 0.995 4.72E-4 0.210 Germany 0.456 1.0 0.026 0.991 1.15E-3 0.218 (1a): Bare kernel (1200 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

(1b): Bare kernel (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

(2a): kernel+buffer+iPyC (1200 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

(2b): kernel+buffer+iPyC (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

(3a): Intact (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

(3b): Intact (1800 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

IAEA CRP-6 Benchmark Fractional Release TRISO Diffusion Release A sensitivity study to examine fission product release from a fuel particle starting with a bare kernel and ending with an irradiated TRISO particle; STORM (Simplified Test of Resuspension Mechanism) test facility Resuspension LACE LA1 and LA3 tests experimentally examined the transport and retention of aerosols through pipes with high speed flow Turbulent Deposition Validation Cases

  • Simple geometry: AHMED, ABCOVE (AB5 & AB6), LACE(LA4),
  • Multi-compartment geometry: VANAM (M3), DEMONA(B3)
  • Deposition: STORM, LACE(LA1, LA3)

Agglomeration Deposition Condensation and Evaporation at surfaces Aerosol Physics

113 MELCOR Modernization Generalized numerical solution engine Hydrodynamics In-vessel damage progression Ex-vessel damage progression Fission product release and transport

MELCOR default radionuclide classes

115 MELCOR default radionuclide classes

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